• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• 한국기계기술학회지
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• 제13권4호
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• pp.85-91
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• 2011

Experimental analysis has been carried out to investigate thermal characteristics of hydraulic system in special vehicles. Hydraulic system performance is largely influenced by oil temperature, and there are considerable performance decline and malfunctions in the system for high temperature conditions caused by heavy load and continuous operation. Transient oil temperature and pressure variation are analyzed and heat generation rates in the several main system parts are compared for various flow rates. With the start of system operation oil temperature gradually increases, and viscosity deceases by about 70% as temperature increases from $20^{\circ}C$ to $80^{\circ}C$. Operation pressure in the hydraulic system decreases with oil temperature, and its variation rate becomes less steep as oil temperature increases. Heat generation rate in hydraulic pump also depends on the oil temperature, and it reaches maximum near $50^{\circ}C$. These results in this study can be applied to optimal design of efficient hydraulic system in special vehicles.

• 항공우주시스템공학회지
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• 제16권3호
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• pp.84-92
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• 2022

전기-정유압식 작동기(EHA)는 독립적으로 유동력원을 운용함에 따라, 복잡한 유압 배관을 제거할 수 있어 누유 및 중량 최소화, 안전성 향상의 장점이 있어 최근 항공기용 비행제어 분야에서 사용되고 있다. 이러한 EHA를 탑재하는 항공기의 경우, 기존 중앙 유압시스템을 탑재한 항공기에 비해 제한된 냉각원에 따른 EHA의 열관리 이슈가 대두된다. 이러한 열관리 이슈의 해결을 위해서는, EHA의 열특성을 예측할 수 있는 열해석 모델이 필요하다. 본 연구에서는 유압펌프 및 전기모터로 구성되는 EHA 유압동력모듈의 내부 회전체를 고압 하에서 고속으로 회전이 가능하도록, 유압동력모듈 내부에 유체 순환 회로를 적용하였다. 적합한 열해석 모델을 구축하고, 유냉식 또는 비유냉식 유압동력모듈 적용에 따른 해석 결과의 비교 및 검토를 통해 EHA의 열특성 영향성을 확인하고자 하였다.

• 드라이브 ㆍ 컨트롤
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• 제18권2호
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• pp.1-8
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• 2021

In this work, a circuit with a hydraulic power unit is formulated as a means of predicting the behavior of the prefill valve in the future. The behavior of the prefill valve can be examined by the measurements of the configured power unit, and the performance is determined by using hydraulic pumps, relief valves, and hydraulic hoses that make up the power unit. In particular, pressure/flow pulsation generated by hydraulic pumps can cause instability in the prefill valve and cause noise-induced degradation of the overall performance and reliability of the hydraulic system containing the prefill valve. Therefore, to study the behavior and performance of the prefill valve in a relatively accurate manner, the prediction of the characteristics of the hydraulic power unit driving the prefill valve is very important. In this study, the pulsation characteristics of the hydraulic pump were analyzed to theoretically demonstrate its relationship with different settings of the power unit, such as relief valve pressure settings and the presence/absence of the hose.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.185-194
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• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• Nuclear Engineering and Technology
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• 제28권6호
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• pp.542-550
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• 1996

Nuclear ship reactors have several features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been peformed under rolling, heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removal to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Journal of Mechanical Science and Technology
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• 제14권9호
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• pp.985-996
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• 2000

The issue of quench is related to safety operation of large-scale superconducting magnet system fabricated by cable-in-conduit conductor. A numerical method is presented to simulate the thermal hydraulic quench characteristics in the superconducting Tokamak magnet system, One-dimensional fluid dynamic equations for supercritical helium and the equation of heat conduction for the conduit are used to describe the thermal hydraulic characteristics in the cable-in-conduit conductor. The high heat transfer approximation between supercritical helium and superconducting strands is taken into account due to strong heating induced flow of supercritical helium. The fully implicit time integration of upwind scheme for finite volume method is utilized to discretize the equations on the staggered mesh. The scheme of a new adaptive mesh is proposed for the moving boundary problem and the time term is discretized by the-implicit scheme. It remarkably reduces the CPU time by local linearization of coefficient and the compressible storage of the large sparse matrix of discretized equations. The discretized equations are solved by the IMSL. The numerical implement is discussed in detail. The validation of this method is demonstrated by comparison of the numerical results with those of the SARUMAN and the QUENCHER and experimental measurements.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.3-8
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• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• 대한기계학회논문집B
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• 제30권9호
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• pp.825-833
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• 2006

To investigate the moderator coolability for CANDU-6 reactors, a test facility (HU-KINS) has been manufactured as a 1/8 scaled-down of a calandria tank. In the design of the test facility, a scaling law was developed in such a way to consider the thermal-hydraulic characteristics of a CANDU-6 moderator. The proposed scaling law takes into consideration of the energy conservation, the dynamic similitude such as dimensionless numbers, Archimedes number (Ar) and Reynolds number (Re), and thermal-hydraulic properties similitude. Using this proposed scaling law, the thermal-hydraulic scaling analyses of similar test facilities such as the SPEL (1/10 scale) and the STERN (1/4 scale), have been identified. As a result, in the case of the SPEL, while the energy conservation is well defined, the similarities of Ar and the heat density are not well considered. As for the similarity of the STERN, while both the energy conservation and the characteristics of Ar are well defined, the heat density is not. In the meanwhile, the HU-KINS test facility with 1/8 length scaled-down is well similitude in compliance with all similarities of the energy conservation, the fluid dynamics and thermal-hydraulic properties. To verify the adequacy of the similarities in terms of thermal-hydraulics, a computational fluid dynamic (CFD) analysis has been conducted using the CFX-5 code. As the results of the CFD analyses, the predicted flow patterns and variation of axial properties inside the calandria tank are well consistant with those of previous studies performed with FLUENT and this implies that the present scaling method is acceptable.